ORAU TEAM

Dose Reconstruction

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Oak Ridge Associated Universities I Dade Moeller & Associates I MJW Corporation

Page 1 of 32

Document Title:

Chapman Valve Manufacturing Company, 

Indian Orchard, Massachusetts

Document Number: ORAUT-TKBS-0033

Revision: 01

Effective Date: 10/16/2006

Type of Document: TBD

Supersedes: 00

Subject Experts: Patricia L. Lee and Cindy W. Bloom

Approval: Signature on File Approval Date: 10/13/2006

Cindy W. Bloom, Document Owner

Approval: Signature on File Approval Date: 10/13/2006

John M. Byrne, Task 3 Manager

Concurrence: Signature on File Concurrence Date: 10/13/2006

Edward F. Maher, Task 5 Manager

Concurrence: James P. Griffin Signature on File for Concurrence Date: 10/13/2006

Kate Kimpan, Project Director

Approval: Signature on File Approval Date: 10/16/2006

James W. Neton, Associate Director for Science

 New Total Rewrite Revision Page Change

FOR DOCUMENTS MARKED AS A TOTAL REWRITE, REVISION, OR PAGE CHANGE, REPLACE THE PRIOR 

REVISION AND DISCARD / DESTROY ALL COPIES OF THE PRIOR REVISION.
Document No. ORAUT-TKBS-0033 Revision No. 01 Effective Date: 10/16/2006 Page 2 of 32

PUBLICATION RECORD

EFFECTIVE 

DATE

REVISION 

NUMBER DESCRIPTION

12/22/2004 00-A New Technical Basis Document: Basis for the Development of an 

Exposure Matrix for Chapman Valve Manufacturing, Orchard, 

Massachusetts. Initiated by Patricia L. Lee.

12/28/2004 00-B Incorporates internal review comments. Initiated by Patricia L. Lee.

02/06/2005 00-C Incorporates OCAS comments. Also addresses new information. 

Extends assumes operational period from 12/31/1948 to 4/30/1949, 

which impacts the assigned intakes, and the 1949 external doses. 

Modified external dose table slightly. Notes the possibility of 

radiological work for Oak Ridge prior to 1948. Initiated by Patricia L. 

Lee.

02/11/2005 00-D Added reference for 1952 newspaper article. Modified wording 

regarding assumed intake dates and penetrating radiation energy 

assumption. Initiated by Cindy W. Bloom.

02/22/2005 00 First approved issue. Initiated by Patricia L. Lee

10/16/2006 01 Approved Revision 01. Replaced boilerplate language in Section 

1.0. Added information from February 2005 Worker Outreach 

meeting. Added additional claim-supplied information. Added 

discussion supporting assumption of unenriched uranium. Added 

information regarding the use of a furnace/incinerator and provided a 

comparison of internal exposures to NUMEC incinerator operators 

with derived Chapman Valve exposures. Added information on the 

radiological safety program from a 1952 newspaper account. Added 

information regarding the month of the previously reported 1948 fire. 

Provided information developed for SEC petition evaluation report. 

Reanalyzed intake regimes as directed by OCAS using the largest of 

the 40 results not associated with the fire, 0.03 mg/L, for each 

sampling period to calculate a non-fire intake rate, and using the 

largest result associated with the fire, 0.08 mg/L, for the June 11, 

1948 result, and keeping the other results the same as for the nonfire scenario to calculate the fire-associated intake scenario. Both 

intake scenarios are directed to be constant distributions in IREP. 

Added a comparison of Chapman derived air concentrations with air 

concentrations measured at other facilities. Added figures showing 

location of chip burner and dust, debris and soil sampling.

Corrected two residual radioactivity external exposure values 

(increased rates by a factor of 2). Incorporated formal internal review 

comments, NIOSH and DOL comments. This revision results in an 

increase in assigned dose and a PER is required. Training required: 

As determined by the Task Manager. Initiated by Cindy W. Bloom.
Document No. ORAUT-TKBS-0033 Revision No. 01 Effective Date: 10/16/2006 Page 3 of 32

TABLE OF CONTENTS

SECTION TITLE

Acronyms and Abbreviations ...............................................................................................................

PAGE

5

1.0 Purpose and Scope .................................................................................................................... 6

2.0 Site Description and Operational History..................................................................................... 7

2.1 Site Description ............................................................................................................7

2.2 Source Term and Process Description .........................................................................8

2.2.1 Incinerator Information .................................................................................... 11

2.3 Safety.........................................................................................................................11

2.3.1 Air Concentrations .......................................................................................... 12

2.3.2 Contamination/Radiation................................................................................. 12

2.4 Incidents.....................................................................................................................12

2.5 Physical Examinations – X-rays..................................................................................13

2.6 Summary Operational Period Assumptions, Workdays, Work Hours, and 

Work Categories.........................................................................................................13

2.7 Cleanup/Residual Contamination Period ....................................................................14

3.0 Estimation of Internal Exposure ................................................................................................ 14

3.1 Uranium......................................................................................................................14

3.1.1 Uranium Bioassay........................................................................................... 15

3.1.2 Uranium Air Sampling ..................................................................................... 16

3.2 Occupational Internal Dose Reconstruction Assumptions and Summary.................... 17

4.0 Estimation of External Exposure ............................................................................................... 18

4.1 Summary of Available Film Badge Data .....................................................................19

4.2 Occupationally Required Medical X-Ray.....................................................................20

4.3 Miscellaneous Information Related to External Dose..................................................21

4.4 Occupational External Dose Reconstruction Assumptions and Summary................... 21

5.0 Estimation of Residual Exposure .............................................................................................. 21

References ........................................................................................................................................ 24

ATTACHMENT A, CHAPMAN VALVE BUILDING LAYOUT AND SOME SAMPLING 

LOCATIONS ..........................................................................................................................28

ATTACHMENT B, COMPARISON OF AIR CONCENTRATIONS CALCULATED FOR 

CHAPMAN VALVE AND AEC MACHINING AND GRINDING OPERATIONS........................ 31
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LIST OF TABLES

TABLE TITLE

3-1 Maximum coworker bioassay data used to estimate intakes ..................................................

PAGE

15

3-2 Inhalation intakes based on coworker data ............................................................................ 16

3-3 Inhalation intake summary for operational period ................................................................... 18

4-1 Maximum film badge results................................................................................................... 20

4-2 External exposure summary................................................................................................... 21

5-1 Radiation/radioactivity levels in August 1991.......................................................................... 22

5-2 Annual internal and external exposure to residual radioactivity .............................................. 23

B-1 Comparison of air concentrations calculated for Chapman Valve and AEC machining 

and grinding operations.......................................................................................................... 32

LIST OF FIGURES

FIGURE TITLE

2-1 Chapman Valve Manufacturing Company ................................................................................

PAGE

8

A-1 Chapman Valve floor plan provided at 2005 Worker Outreach Meeting ................................. 28

A-2 Locations of dust and debris samples collected in 1992 in or near Building 23....................... 29

A-3 Locations of verification samples collected in 1996 in Building 23.......................................... 30
Document No. ORAUT-TKBS-0033 Revision No. 01 Effective Date: 10/16/2006 Page 5 of 32

ACRONYMS AND ABBREVIATIONS

AEC U.S. Atomic Energy Commission

AWE atomic weapons employer

cm centimeter

d day

dpm disintegrations per minute

DOE U.S. Department of Energy

EEOICPA Energy Employees Occupational Illness Compensation Program Act of 2000

ft foot

FUSRAP Formerly Utilized Sites Remedial Action Program

GM Geiger-Mueller

GSD geometric standard deviation

HASL Health and Safety Laboratory

hr hour

ICRP International Commission on Radiological Protection

in. inch

IREP Interactive RadioEpidemiological Program

keV kiloelectron-volt, 1,000 electron-volts

L liter

m meter

MDA minimum detectable activity

mg milligram

mR milliroentgen

mrad millirad

mrem millirem

NYDO New York Department of Operations

NYOO New York Operations Office

pCi picocurie

R roentgen 

SF source and fissionable

U.S.C. United States Code

μg microgram

μR microroentgen

§ section or sections
Document No. ORAUT-TKBS-0033 Revision No. 01 Effective Date: 10/16/2006 Page 6 of 32

1.0

Technical basis documents and site profile documents are not official determinations made by the 

National Institute for Occupational Safety and Health (NIOSH) but are rather general working 

documents that provide historic background information and guidance to assist in the preparation of 

dose reconstructions for particular sites or categories of sites. They will be revised in the event 

additional relevant information is obtained about the affected site(s). These documents may be used 

to assist NIOSH staff in the completion of the individual work required for each dose reconstruction.

PURPOSE AND SCOPE

In this document the word “facility” is used as a general term for an area, building, or group of 

buildings that served a specific purpose at a site. It does not necessarily connote an “atomic weapons 

employer facility” (AWE facility) or a “Department of Energy facility” (DOE facility) as defined in the 

Energy Employees Occupational Illness Compensation Program Act of 2000 [EEOICPA; 42 U.S.C. § 

7384I(5) and (12)]. EEOICPA, as amended, provides for employees who worked at an AWE facility 

during the contract period and/or during the residual contamination period.

Employment at an AWE facility is categorized as either (1) during the contract period (i.e., when the 

AWE was processing or producing material that emitted radiation and was used in the production of 

an atomic weapon), or (2) during the residual contamination period (i.e., periods during which NIOSH 

has determined there is the potential for significant residual contamination outside of the period in 

which weapons-related production occurred). For contract period employment, all occupationally 

derived radiation exposures at the facility must be included in dose reconstructions. NIOSH does not 

consider the following exposures to be occupationally derived:

• radiation from naturally occurring radon present in conventional structures

• radiation from diagnostic X-rays received in the treatment of work-related injuries

For residual contamination period employment, only the radiation exposures defined in 42 U.S.C. 

§ 7384n(c)(4) (i.e., radiation doses received from DOE/AEC-related work) must be included in dose 

reconstructions. Radiation dose received from DOE/AEC-related work includes: (1) radiation from 

radon consistent with NIOSH’s policies for including such radiation in the contract period; and (2) 

medical screening X-rays, but not diagnostic X-rays for the treatment of work-related injuries. It 

should be noted that: (1) under subparagraph A of 42 U.S.C. § 7384n(c)(4), radiation associated with 

the Naval Nuclear Propulsion Program is specifically excluded from the employee’s radiation dose; 

and, (2) under subparagraph B of this section, radiation from a source not covered by subparagraph A 

that cannot be reliably distinguished from radiation that is covered by subparagraph A is considered 

part of the employee’s radiation dose. This site profile covers only exposures resulting from nuclear 

weapons-related work. Exposures resulting from non-weapons-related work, if applicable, will be 

covered elsewhere.

This site profile provides an exposure matrix for workers at the facility listed as Chapman Valve 

Manufacturing Company in Indian Orchard, Massachusetts. Chapman Valve was a manufacturer of 

nonradioactive valves and manifolds, some of which were purchased by the Federal government. In 

1948, Chapman Valve machined uranium rods for Brookhaven National Laboratory. In 1959, 

Chapman Valve was purchased by the Crane Company.

Section 2.0 of this document describes the Chapman Valve site and its history including some 

information about the radiological processes and source terms as well as the radiological controls and 

monitoring practices. Sections 3.0 and 4.0 discuss internal and external dose, respectively. Section 

5.0 provides information for use in reconstructing dose during the residual period.
Document No. ORAUT-TKBS-0033 Revision No. 01 Effective Date: 10/16/2006 Page 7 of 32

2.0

The information that follows supports an assumed period of AEC operations at Chapman Valve from 

January 1, 1948, through April 30, 1949, involving AEC-contracted uranium work. This analysis 

assumed that the residual contamination period extended from May 1, 1949, through December 31, 

1993. Exposures during the decontamination period in 1994 and 1995 are not addressed.

SITE DESCRIPTION AND OPERATIONAL HISTORY

A 1952 newspaper article (Springfield Union News ca. 1952) stated that a small group of Chapman 

Valve employees might have been involved in work with radioactive materials for Oak Ridge prior 

to1948. At this time, no data have been found to substantiate or refute this statement. Additional 

information is being sought, and the site profile will be updated if necessary to address an earlier 

exposure period. Many reports indicate that Chapman Valve was involved in the manufacture of 

nonradioactive equipment and parts for the Manhattan Engineer District.

The Chapman Valve radiological source term consisted primarily of natural uranium metal, uranium 

oxides, and natural uranium’s short-lived progeny. Long-lived progeny in the uranium series prevent 

significant ingrowth past 234U in the 238U decay series and beyond 231Th in the 235U decay series.

2.1 SITE DESCRIPTION

The Chapman Valve main office and works were located on Hampshire Street in Indian Orchard, 

Massachusetts. Uranium shipments were made to Oak Street and claims refer to uranium operations 

in the building on Pine Vale Street, which this document assumes to be Building 23. Figure 2-1 

shows the location of Building 23 on the site.
Document No. ORAUT-TKBS-0033 Revision No. 01 Effective Date: 10/16/2006 Page 8 of 32

Figure 2-1. Chapman Valve Manufacturing Company (ORNL 1997).

A former employee recalls that in 1947 Chapman Valve set aside approximately one-third of 

Department No. 40 at the Chapman Valve site for the machining of uranium rods for Brookhaven 

National Laboratory (Fiore 1987, Attachment). It is not clear that the Department No. 40 designation 

would be an indicator of whether or not someone worked in the restricted area of Building 23. The 

“set aside” portion of the building measured approximately 200 ft long by 60 ft wide and greater than 

50 ft high. The area was separated from the remainder of the building by a floor-to-ceiling wooden 

partition, which has since been removed. Although there is no definitive statement regarding where 

the restricted area was in the building, based on later radiological surveys, it is likely that it was 

located in the western portion of the building. The building that contained the uranium operations has 

been vacant since Crane discontinued all manufacturing at Indian Orchard early in 1987.

2.2 SOURCE TERM AND PROCESS DESCRIPTION 

Under contract with Brookhaven National Laboratory, Chapman Valve machined uranium metal 

during the period January to November 1948 (Young 1987). Preparation for this effort may have 

begun in November 1947 with the initiation of health and safety programs (Wolf 1947). An inventory 

report indicated that Chapman Valve had less than 50 pounds of uranium as of January 1948 

(Morgan 1948). Actual production may not have started until May 1948, but a set of contamination 

measurements from March 19, 1948, indicates start-up activities may have occurred prior to May 

1948 (LeVine 1948). Records indicate that machining operations ceased in November 1948, and it 
Document No. ORAUT-TKBS-0033 Revision No. 01 Effective Date: 10/16/2006 Page 9 of 32

was assumed that all identified uranium materials and scrap were sent off site before January 1, 1949 

(Fiore 1987).

The AEC Medical Director from the New York office (Wolf 1947) stated: “On November 24 [1947], a 

conference was held with Chapman Valve and Ferguson personnel as a result of which a complete 

set of health and safety recommendations were made a few days later in writing. Chapman is 

planning to undertake work involving the machining of uranium.” 

Twenty-six tons of uranium rods were sent on January 9, 1948, to the Chapman Valve Oak Street 

facility (Fiore 1987). There is a single report (LeVine 1948) of three radioactive dust samples dated 

March 19, 1948, but production-scale (versus trial run) machining may not have started until May 

1948, as indicated in an attachment to the Fiore (1987) letter. In January 1948, Wolf (1948a) begins

the notation regarding Chapman Valve with the words “A survey of proposed uranium machining 

operation” indicating that the facility was not machining large quantities of uranium. In May 1948, 

Wolf (1948b) stated “Health and safety preparations at the Chapman Valve plant in Indian Orchard, 

Mass. for the proposed machining operation were investigated by the Radiation Survey and Safety 

Sections,” indicating the facility had not begun machining as of May 6, 1948.

An inventory form (Morgan 1948, 1949) indicated that the beginning inventory in January 1948 was 

about 48 [pounds] of “SF Material” (source and fissionable material, assumed to be uranium). From 

January to July 1948, documentation indicated that about 141,200 pounds of SF material were 

received and about 42,600 pounds in the form of slugs and turnings were shipped off site. Because 

that same inventory form was attached to Morgan’s (1949) letter, it was unclear if another shipment of 

SF was made. Uranium rods were brought to the facility by railroad on a track immediately adjacent 

to the building. Rods were cut by mechanical saw and then machined to the desired shape.

Records regarding the Chapman Valve AEC machining work indicate the uranium was of normal 

enrichment (normal enrichment is about 0.72% U-235 weight). Chapman Valve uranium processing 

was related to reactor needs, and the use of enriched uranium in reactors was rare until 1950. The 

Chapman Valve contract with Brookhaven National Laboratory was to machine uranium for reactor 

pile, which was likely the Brookhaven Graphite Research Reactor. This breeder reactor first went 

critical in 1950 and was initially loaded with natural uranium (BNL 2006).

Chapman Valve participants in the February 2005 Worker Outreach Program meeting noted a 

contamination result of 2.16% enriched uranium was reported in later years (ORAUT 2005a). A 

review of the ORNL (1992) document that reported this result shows that two dust and debris 

samples, M10 and M31, were analyzed isotopically (see Attachment A, Figure A-2, for locations). 

Reported U-238 results were respectively 25 ±7 and 120 ±10 pCi/g of dust and debris (2-sigma 

counting errors). Sample M31, which was collected to the north of a west door of Building 23 (see 

Figures in Attachment A) was reported as enriched to 2.16% U-235 by weight. The report did not 

include U-235 results or mention the uncertainty associated with the enrichment determination. A 

second report (ORNL 1997) included seven radioisotopic measurements of soil at Chapman Valve 

(see Figure A-3 for locations) and attributed the measured results to background levels. Soil sample 

VS1 from the 1997 survey was collected just north of the location where a chip burner was shown on 

a 1996 map and perhaps slightly south of the area where M31 had previously been collected. The 

results were 14 ±1 pCi/g U-238 and 1 ±0.07 pCi/g U-235, which are consistent with unenriched 

uranium.

A number of factors, in addition to counting techniques, add to the uncertainty of isotopic 

measurements, and it appears that the uncertainties reported in ORNL (1992, 1997) only include 

consideration of counting errors. In addition, if gamma spectrometry were used to analyze the dust 
Document No. ORAUT-TKBS-0033 Revision No. 01 Effective Date: 10/16/2006 Page 10 of 32

and debris sample, and if the 186-keV photons [the most abundant photons for both U-235 and Ra226 (ORNL 2003)] were used to quantify both U-235 and Ra-226, there is a possibility that some Ra226 was inadvertently reported as U-235. 

Given that there are no other data to indicate that the uranium processed at Chapman Valve was 

enriched, and that enriched uranium was not typically sent to non-government-controlled sites without 

extra precautions, it is concluded that this single 1992 sample result is not evidence of enriched 

uranium work at Chapman Valve in 1948.

Crane Company (Young 1987, Enclosure 1) noted, “One AEC memorandum, William to Kelly, date 

illegible, indicates that Chapman Valve may also have conducted rolling operations on uranium 

metal.” A review of available references (the William memorandum was not available) and claimantprovided information does not indicate that Chapman Valve ever rolled uranium rods. An October 

1948 New York Directed Operations report (AEC 1948a) does not include Chapman Valve in the 

mentioned uranium rolling sites.

In July 1992, ORNL (1992) states:

After contract work was completed, Chapman Valve had in their possession over 

27,000 pounds of metal scrap, oxides, and sweepings. Termination of these 

operations is indicated in a Chapman Valve letter dated November 8, 1948, which 

requested termination of AEC film badge services. All radioactive residues and 

contaminated materials were surveyed by Brookhaven Medical Group and shipped offsite. The actual shipment date is unknown, but the shipment probably took place in 

December 1948.

This offsite shipment date appears to be based on a Crane Company letter, which noted that there 

was correspondence that indicated that all radioactive residues and contaminated material were 

shipped off site in December 1948. Another note indicates that the 27,000 pounds of metal scrap, 

oxides, sweepings, etc. were shipped off site several months after the contract was complete. 

Morgan (1949) is unclear regarding whether another shipment of SF material would be made after 

January 25, 1949. ORNL (1992) and an enclosure to Fiore (1987) state that Chapman Valve had in 

its possession over 27,000 pounds of metal scrap, oxides, sweeping, etc., for several months beyond 

completion of the contract. Both documents refer to a Chapman Valve letter dated November 8, 

1948, that requested termination of AEC film badge services and indicated that all radioactive 

residues were surveyed by the Brookhaven Medical Group and shipped off site in December 1948. 

An Electro Metallurgical (1949) weekly production report for April 1 to 30, 1949, mentioned about 

28,000 pounds of metal received from Chapman Valve, which indicated that the final shipment would 

have been no later than April 30, 1949.

It was noted that 100 film badges were needed for the film badge program (Musgrave 1948). The film 

badge records show that workers were employed by Chapman Valve, H. K. Ferguson, or Brookhaven 

National Laboratories. The records indicate that at least for one week in May, two shifts may have 

been involved in the AEC work (AEC 1948b). The available bioassay records include 37 names. Job 

categories noted in the records included supervisory personnel (electrician, heat treater, steamfitting 

and plumbing, machine shop and maintenance, inspector, guard, personnel), engineer, guards, 

inspectors, trades (electrician, machine repair, master mechanic, milling machine operator, steamfitter 

and turret lathe operator), helpers (janitor, weigher, not specified), and several assistants to the 

director of research. The external dose data sheets also list jobs such as brushing and packer. A 

quick review of claim information indicates that job titles listed in the claims might not always be 

consistent with the job titles held in 1948.
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Area access controls were in place at Chapman Valve, but these controls are not clearly associated 

with job categories. Rather, they were an issue of both security and contamination control. A May 26, 

1948, handwritten, two-page, “Plant Protection” list might indicate the number of hours in a week that 

workers were in the AEC work area (AEC 1948b, pp. 155–157). The reported number of hours 

ranged from zero to 58; the majority of entries were less than 40 hours. Eleven entries were greater 

than 40 hours. Similar lists including dates from May 3, 1948, through the week of May 19, indicate 

that these lists show the number of times and the number of hours a [film] badge was worn.

This document assumes that workers might have been exposed to uranium for up to 8 hours per day, 

250 days per year, although it appears likely that uranium work was not in full swing until May 1948 

and was essentially over in November 1948.

2.2.1

A furnace was used during the 1948 AEC operations, probably for oxidizing uranium turnings and 

scraps to reduce the risk for uranium fires during storage and shipping. The Chapman Valve Final 

Remedial Action Report (Robbins 1996) and the February 14, 2005, Chapman Valve Worker 

Outreach Program meeting discussion included the terms chip incinerator, chip burner, and cracking 

furnace (ORAUT 2005a). On the map (Author unknown ca. 1996) provided at the meeting and shown 

in Figure A-1, a location for a chip burner is marked in the west side of Building 23 in an area 

consistent with the AEC restricted area. At what appears to be a similar location (see Figure A-2), 

ORNL (1992) reported elevated contamination near a window. Robbins (1996) reported 

contamination was found in the area “where a chip burner was located in the southwest corner of Grid 

A-1 [of Building 23] that exhausted to the atmosphere out a nearby window. The exhaust location and 

the shape of the roof of the building would lead to the deposition of more contamination on the south 

roof than the north roof as indicated by characterization measurements.” On May 4, 1948, the AEC

(1948c) made air concentration measurements at the furnace outlet on the roof and on the roof 4 feet 

from the outlet.

Incinerator Information 

2.3 SAFETY 

An affidavit from a former Chapman Valve employee describes some of the radiological safety 

controls (Redacted 1987, Attachment 1). Workers would remove their clothes and don white coveralls 

prior to entering the uranium manufacturing area. Workers wore dosimeters on their uniforms. 

Inspectors would pass through the site carrying Geiger counters. At the completion of a day’s work, 

the employees would return to the dressing room and remove their white coveralls. Each employee 

was then required to take a shower as a safety measure. The floors were swept every night. Cuttings 

from the machining process were stored in drums and disposed of periodically by Brookhaven. A 

later description of activities by this former tinshop employee (Claim File redacted, DOL Initial

Case_redacted.pdf, p. 31-32), noted that he “entered the work area to perform jobs... which consisted 

of taking measurements of splash guards, chip deflectors, suction systems, holding pans or [taking 

measurements of] anything else required. We then returned to our shop, performed our layouts and 

fabricate[d] these articles (some took a week) returned ASAP to install in place. Periodically we had 

to clean the main suction system, which we had installed, as it was highly susceptible to catching fire, 

it did on occasion (it’s a matter of record). [My assistant] and I were never required (to my knowlede 

[sic]) to put on a uniform, probably because of the short time spent on site. The long jobs wer [sic} 

done on weekends when the operation was down.”

A 1952 newspaper article (Springfield Union News ca. 1952) included the following description of the

radiation protection program:
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Workers in this location were fully protected by means of special clothing and entered 

the workshop only through a labyrinth entrance into a specially constructed room 

where they changed their street clothes for the special uniforms. Each worker was 

equipped with a badge which he wore on his chest at all times while at work. These 

badges contained a sensitive film, which was frequently developed to reveal if the

worker had received any dangerous radiation and also the degree of absorption. 

Geiger-Muller and other sensitive instruments kept a close check on the safety of 

workers in areas where radiation was most likely to occur. Many of the key personnel, 

who were in and out, carried a fountain-pen-like instrument known as a minometer 

[type of personal direct reading dosimeter], which was checked daily as an infallible 

safety precaution. 

Eisenbud (1948) indicated that employees could use the lunchroom for snacks or smoking provided 

the hands and face were thoroughly washed and clean laboratory coats and shoe covers were worn 

over protective clothing. His requirement that the lunchroom be surveyed and decontaminated, if 

contamination was detected, implies the expectation that these procedures might not eliminate the 

spread of contamination to the lunchroom and the locker room. Records of surface contamination 

surveys in these areas have not been located. Uranium air concentration measurements indicate 

elevated activity in the lunch and wash rooms.

2.3.1

During World War II, permissible levels for uranium dust in air were set at 500 μg/m3 for insoluble 

uranium compounds and 150 μg/m3 for soluble uranium compounds. After the war, the University of 

Rochester lowered its recommendation for soluble uranium compounds to 50 μg/m3 based on the 

chemical toxicity, which for natural uranium is equivalent to 70 dpm/m3

. This level was based 

primarily on animal studies. The Medical Division of the AEC New York Operations Office (NYOO) 

felt that a maximum permissible level was unknown and should be based on human data. Therefore, 

the 50 μg/m3 level was referred to as the preferred level (AEC 1949).

Air Concentrations 

Very few records of uranium air concentration measurements taken during AEC operations were 

found for Chapman Valve. Results included data sets from May 4, 1948, and May 24, 1948 (AEC 

1948c, d). 

2.3.2

Available documentation of radioactive contamination during the operational period was limited to 

three wipe samples from March 19, 1948, one of which could not be evaluated. The results were 

8 counts per minute for the cracking furnace and 300 counts per minute for the “centerless grinder.” 

No measurements of radiation levels for the AEC uranium operational period were located.

Contamination/Radiation

2.4 INCIDENTS 

Chapman Valve (Fox (1949) notes, “A fire occurred in the restricted area of the AEC project and on 

June 11th [1948] urine for analyses... was sent to the School of Medicine and Dentistry of Rochester 

New York.” Collection of urine samples from the workers who put out the fire and from cleanup 

personnel suggests that the uranium was involved in the fire. The exact date of the 1948 fire has not 

been found, although AEC’s transmittal note (Wolf 1949) regarding urinalysis results for samples 

“collected from the seven employees involved in the fire fighting episode last June [1948]” is evidence 

that the firefighting occurred in June. This appears to be the fire referred to in claim redacted. No 

documentation was found on the exact date of the fire.
Document No. ORAUT-TKBS-0033 Revision No. 01 Effective Date: 10/16/2006 Page 13 of 32

A claimant provided a description of an explosion:

There was a significant event at Chapman Valve sometime during those two years 

[1948-1949]. It appears that a cupola blew up on the first shift destroying a large 

portion of the plant with the result being that one worker was killed and several were 

seriously injured. The plant was closed for some time after this incident. The resultant 

fire consumed three floors of one section of the building.

No evidence could be found to support a major fire during AEC uranium operations at Chapman 

Valve, except as noted for June 1948. However, a newspaper search revealed that an explosion took 

place when power to an oil transformer was engaged in the early 1940s. Seven workers were injured, 

and a small brick building that housed the electrical equipment collapsed. The newspaper (Springfield 

Union News ca. 1942) reported 17 hours of lost production time and that:

pouring operations in the iron and steel foundries which turn out giant valves for the 

navy [sic] continued without interruption throughout the night, Duggan said, so that a 

backlog of work would be on hand for the resumption of finishing operation in the 

machine shop.

The Chapman Valve Special Exposure Cohort petition stated that documentation at other facilities 

shows that uranium fires are prevalent in uranium milling/lathing operations due to the fact that 

uranium is a pyrophoric material (ORAUT 2006). Although fires were not rare at uranium metal 

processing sites in the early days of Manhattan Engineer District and AEC operations, more rigorous 

controls were being employed to prevent fires, as seems to be the case at Chapman Valve after the 

June fire. One claimant noted his job was to clean the suction system to reduce the possibility for 

fires (Claim File redacted, DOL Initial Case_redacted.pdf, p. 31-32). No other claims were found that 

mentioned fire. A comment from the February 14, 2005 Worker Outreach Program meeting (ORAUT 

2005a) noted, “If they were machining uranium on the lathe, there would have been heat and fire.” 

Based on general knowledge of uranium metal operations during this early period, it is conceivable

that one or more additional small fires occurred. Such fires if they occurred were probably controlled 

by the local workforce.

2.5 PHYSICAL EXAMINATIONS – X-RAYS 

No information regarding AEC-required physical examinations for Chapman Valve employees has 

been located.

2.6 SUMMARY OPERATIONAL PERIOD ASSUMPTIONS, WORKDAYS, WORK HOURS, 

AND WORK CATEGORIES 

Because bioassay data and film badge data were used to estimate internal and external exposures 

during the AEC uranium operations, an estimate of workdays or work-hours per year is not important.

For later years, it was assumed that workers worked 8 hours per day for 50 weeks per year, for a total 

of 2000 hours per year.

While different tasks resulted in differences in exposures, it is evident from the records that workers 

did not always perform the same tasks. Exposure assignments are based on data that are suggestive 

of workers’ exposures and further modified by uncertainty parameters, when appropriate, to ensure 

that the reconstructed dose distributions capture the larger exposures. No attempt has been made to 
Document No. ORAUT-TKBS-0033 Revision No. 01 Effective Date: 10/16/2006 Page 14 of 32

sort workers into exposure categories. Depending on the organ of interest and the ancillary data 

associated with a specific claim, additional considerations might be appropriate.

2.7 CLEANUP/RESIDUAL CONTAMINATION PERIOD 

A former tinshop employee (Claim File redacted, DOL Initial Case_redacted.pdf, p. 31-32) recalled 

disassembling all sheet metal fabrications at the end of operations and removing them to another area 

for disposal. He also stated that he helped with “washing down the wall and the entire Dept., 

including removal of the floor bricks, (contrary to some who said they were never removed) removing 

6” of concrete flooring with re-enforced wiring and hauling away for proper disposal.” He further noted 

that the Carpenter shop installed new concrete and flooring in the area, and that he worked in 

Welding and Fabrication, which was in that same area from 1966 to January 1, 1983.

Crane discontinued plant operations in June 1986 and was in the process of closing down the plant 

for sale as of August 1987 (Young 1987). ORNL (1992, 1997) stated, “The building that contained the 

uranium operations has been vacant since Crane discontinued all manufacturing at Indian Orchard in 

1987.” However, information in the claim files (redacted) indicates that the building was not closed 

until February 22, 1991, or later. This document assumes that residual exposures occurred from May 

1, 1949 through December 31, 1993. Site decontamination took place in 1994 and 1995. This period 

of exposure is not addressed in this document.

No documentation has been found to indicate the radiological condition of the site when operations 

ceased and little information is available to describing cleanup activities for this period. Redacted

(1987) indicates that records that might have contained information related to radiological conditions 

were destroyed during the 1970s. However, elevated levels of contamination were found in surveys 

performed in the early 1990s (ORNL 1992) under the Formerly Utilized Sites Remedial Action 

Program (FUSRAP). This period is discussed further in Section 5.0.

3.0

The primary source of internal radiation exposure at Chapman Valve was uranium dust produced from 

the manipulation and oxidation of the metals during machining and related processes. It is assumed 

that the uranium was of natural enrichment. Recycled uranium did not enter process streams until 

1952, so no recycled uranium would have been processed at Chapman Valve. 

ESTIMATION OF INTERNAL EXPOSURE

3.1 URANIUM

Human and animal studies have indicated that oxides of uranium can be very insoluble (ICRP 1995), 

indicating absorption type S (0.1% and 99.9% with clearance half-times on the order of 10 minutes 

and 7000 days, respectively). Other in vitro dissolution studies of compounds found at uranium 

facilities have shown that oxides of uranium exhibit moderate solubility (Eidson 1994; Heffernan et al. 

2001) suggesting absorption type M (10% and 90% with clearance half-times on the order of 10 

minutes and 140 days, respectively). In vitro dissolution tests on oxides produced from uranium metal 

during depleted uranium armor penetrator tests have indicated multicomponent dissolution rates, with 

25% of uranium dissolving with a half-time of less than or equal to 0.14 days and 75% dissolving with 

a half-time of 180 days. Because there was no specific information on the solubility of aerosols 

produced during operations, this analysis assumed that both types M and S were available. The 

selection of absorption type should depend on the organ of interest.
Document No. ORAUT-TKBS-0033 Revision No. 01 Effective Date: 10/16/2006 Page 15 of 32

3.1.1

Individual uranium urinalysis data are available for some workers at Chapman Valve. Results less 

than 0.01 mg/L were reported as zero. A note on a sheet of uranium results (Author unknown 1948) 

states, “The uranium content of those samples listed as containing less than 0.01 mg U/l is below the 

limit for reliable determination by the photofluorometric method.” The urinalyses range from <0.01 to 

0.08 mg/L. For unmonitored workers or unmonitored periods, this document analyzes the bioassay 

results to provide estimates of coworkers’ uranium intakes.

Uranium Bioassay

A total of 40 uranium urinalysis results are available for Chapman Valve employees. The first 

available bioassay samples, dated June 11, 1948, were analyzed by the School of Medicine and 

Dentistry of Rochester, New York (Fox 1949) and were collected because of a fire involving uranium. 

Two workers who put out the fire had results of <0.01 mg/L. The five workers, who were involved in 

cleanup, had results ranging from 0.01 mg/L to 0.08 mg/L. The actual date of the fire is unknown, 

although AEC (Wolf 1949) noted in the transmittal of the urinalysis results that the samples had been 

“collected from the seven employees involved in the fire fighting episode last June [1948].” Only two 

of the workers involved in putting out the fire or in the associated cleanup had later results. Bioassay 

results were also available for urine samples collected on July 26-27, September 8-9 and October 7 of 

1948 from 22, 6 and 5 workers, respectively.

To calculate the intake not associated with the fire, the largest of the 40 results not associated with 

the fire, 0.03 mg/L, was assigned to each sampling period (the period when results were collected due 

to the fire was also assigned a bioassay result of 0.03 mg/L). To calculate intakes that include the fire 

scenario, the largest result associated with the fire, 0.08 mg/L, was used for the June 11, 1948, result, 

and results for the other sample dates were again assumed to be 0.03 mg/L. The daily uranium 

excretion in urine was calculated by multiplying the results in mg/L by the specific activity of natural 

uranium (682.96 pCi/mg) and by reference man’s daily urine output (1.4 L/day) (ICRP 1975). Table 3-

1 shows the bioassay values used to estimate upper bound inhalation intakes based on an acute 

exposure in June 1948 and a chronic exposure from January 1, 1948, to April 30, 1949.

Table 3-1. Maximum coworker bioassay data used to estimate intakes. 

Bioassay date

Fire scenario 

(pCi/day)

No fire scenario 

(pCi/day)

6/11/1948 76.5 28.7

7/27/1948 28.7 28.7

9/8/1948 28.7 28.7

10/7/1948 28.7 28.7

When intakes are estimated from bioassay data, the mode of intake is usually assumed to be 

inhalation, unless there is information that indicates that other modes of intake are more likely. When 

using bioassay data, the inhalation intake model assumes that some of the intake behaves as 

ingested material. In general, intakes from bioassay will be larger when an inhalation rather than an 

ingestion intake is assumed.

Although the exact date of the June 1948 fire is unknown, it seems reasonable to assume that it could 

have occurred as early as June 1, 1948, and this should be considered when fitting individual 

bioassay data. For the coworker data, assuming the intake from the fire occurred on June 10, 1948 

resulted in the largest total intake for the period January 1, 1948 to April 30, 1949, and provided a 

satisfactory fit to the data; the other assumed June fire dates produced total intakes that were within 

3.6% of this largest intake. The intake from the fire might have occurred over several days, but it was 

assumed to be an acute intake when fitting the data to simplify assumptions. The intakes were 
Document No. ORAUT-TKBS-0033 Revision No. 01 Effective Date: 10/16/2006 Page 16 of 32

calculated with IMBA ExpertTM ORAUT-Edition, Version 4.0.9, assuming an absolute uniform error of 

1 and a normal error distribution. To calculate the level of chronic intake for workers, who might not 

have been on site during the June fire, the June 11, 1948, bioassay was set to the maximum result 

not associated with the fire.

As can be seen from the total intake results in Table 3-2, the accounting for the acute intake (as might 

occur from fires) with a chronic intake assumption results in intake determinations that are sufficiently

large to account for possible acute intake situations that occurred over shorter periods. This is 

evidenced by the similar results for total intakes from fire and non-fire scenarios obtained in this 

analysis.

 Table 3-2. Inhalation intakes based on coworker data. 

Scenario

Absorption 

type

Chronic

1/1/1948 to 4/30/1949

(pCi/d)

Acute 6/10/1948

(pCi)

Total

(pCi)

U machining and fire M 4.81E+02 2.17E+03 2.35E+05

U machining and fire S 1.43E+04 7.39E+04 7.00E+06

U machining M 4.96E+02 Not applicable 2.40E+05

U machining S 1.49E+04 Not applicable 7.23E+06

The Chapman Valve chronic inhalation rates for 1948 through April 30, 1949 in Table 3-2 are, 

respectively, two and more than four times the types M and S median Y-12 coworker inhalation 

intake rates for the period 1947 through 1952 (ORAUT 2005b). The intake scenarios to be used for 

Chapman Valve dose reconstruction are summarized in Table 3-3.

3.1.2

The AEC provided analyses for air samples collected on May 4 and 24, 1948 at Chapman Valve (AEC

1948c, d). The results from May 4 (AEC 1948c) are measurements of the effluent from the furnace 

outlet on the roof and are not appropriate for determining worker doses (it is unlikely that any worker 

spent significant time 4 ft from a rooftop furnace outlet). The results from May 24, 1948 (AEC 1948d) 

are 10-minute grab samples listed as “Inspection Bench, Packing Bench, Work Bench, Wash Room 

and Lunch Room.” Because it is not clear how these samples relate to the workers’ activities, they 

are not directly used in determining worker intakes. However, an inhalation intake can be calculated 

by assuming that the maximum result reported, 29.1 dpm alpha/m3 measured at the workbench, was 

the concentration for the entire work year as follows. 

Uranium Air Sampling

An intake in pCi was calculated by dividing the maximum reported workplace air concentration by 

2.22 dpm/pCi and multiplying this result by the annual breathing rate, 2,000 hours per year, and 

1.33 years. The breathing rate of 1.2 m3

/hr was used based on the default for light work (ICRP 1994, 

Table 6, p. 23). The result is an inhalation intake of 4.18 × 104 pCi, as compared to the total values in 

Table 3-2. Intakes from resuspended material and ingested material might increase this number by 

about 10% based on similar calculations in other site profile documents. Based on reported beta film 

badge results, it is likely that the uranium workload increased during June through September 1948, 

which would likely have increased air concentrations, so it would be reasonable to assume a factor of 

2 to 4 increase in time-weighted air concentrations during the months of June through September, 

because of the change in workload throughout the year. Because there are more bioassay than air 

sample results and the bioassay results are spread throughout the work year and are likely to be a 

better indicator of worker exposures, especially during the production period, the air sample data are 

not used to calculate intakes at Chapman Valve.
Document No. ORAUT-TKBS-0033 Revision No. 01 Effective Date: 10/16/2006 Page 17 of 32

During the February 14, 2005, Worker Outreach Program meeting attendees provided air 

concentrations and bioassay data collected from workers involved in 1966 and 1967 NUMEC

incinerator operations. The incinerator operation was referred to as “primitive” by NUMEC (Schnell 

and Caldwell 1967), and was thought to be perhaps a comparable operation. Although NUMEC 

appears to have handled greater quantities and higher enrichments of uranium, and 1960s NUMEC 

operations might have been quite different from the 1948 Chapman Valve work, comparisons can be 

made. The average breathing zone concentration at different NUMEC locations ranged from 848 to 

6,666 pCi/m3 (these could be a factor of 2 to 100 lower when normalized for the difference in specific 

activity between the NUMEC and Chapman Valve uranium enrichments); these compare with 

calculated exposures (not considering the fire scenario) of 75 to 2,270 pCi/m3 derived from Chapman 

Valve coworker bioassay results. The bioassay data for workers in the incinerator area at NUMEC 

were also available in the provided report and these, too, were considered. The 1966 and 1967 

NUMEC incinerator operators’ uranium urinalysis results (Schnell and Caldwell 1967) were based on 

activity measurement and because NUMEC handled a variety of uranium enrichments this is 

considered in the comparisons. The NUMEC incinerator operators’ urinalyses in terms of mass 

concentrations would have ranged from 37 to 68 μg/L for normal uranium, 16 to 29 μg/L for 2% 

enriched uranium, and 0.24 to 0.45 μg/L for 93.5% enriched uranium. Bioassay results at Chapman 

Valve ranged from <10 μg/L to 80 μg/L, and the maximum result of 30 μg/L (not considering results 

collected because of the June fire) was used to estimate intakes.

Air concentrations were measured on May 4, 1948 at the furnace outlet on the roof. The four 

measurements ranged from 1,010 to 29,100 dpm/m3

. Two more measurements made the same day 

4 ft from the roof exhaust were 136 and 2,280 dpm/m3

. If workers breathed 1% of the maximum 

furnace exhaust air concentration, which was 29,100 dpm/m3 for 2,000 hours per year, the intake rate 

would be 862 pCi/d in comparison to the range of chronic intake calculated from bioassay, which is 

481 to 14,900 pCi/d. Based on the description of Chapman Valve’s uranium processing, which 

indicates that the work was done in less than a year and that the furnace use was unlikely to have 

been continuous for uranium processing, it seems very unlikely that any worker was exposed to the 

maximum air concentration from the furnace exhaust for 2,000 hours per year, and direct exposure to 

the undiluted effluent would have been highly unlikely because it was an incinerator exhaust.

A comparison of the air concentrations derived from the Chapman Valve intakes in Table 3-3 and air 

concentration data collected by AEC for typical uranium operations related to machining is shown in 

Attachment B.

3.2 OCCUPATIONAL INTERNAL DOSE RECONSTRUCTION ASSUMPTIONS AND 

SUMMARY

The uranium photofluorometry urinalysis reporting limit at Chapman Valve was 0.01 mg/L. Uranium 

oxides could be either absorption type M or S.

The assumed operational exposure period ran from January 1, 1948, to April 30, 1949, which this 

analysis assumes to be the uranium intake period. A uranium fire is assumed to have occurred 

sometime between June 1 and 11, 1948. The intake from the fire can be assumed to be acute and to 

have occurred on June 1, 1948, if no other information is available.

For unmonitored workers or unmonitored periods, Table 3-3 lists inhalation intakes for dose 

reconstruction. Chronic intakes are given in units of pCi per calendar day and acute intakes are given 

in pCi. Four different intake scenarios are listed. The first or second scenario accounts for exposure 

to the uranium fire in June 1948. The third or fourth scenario can be used when an individual had no 
Document No. ORAUT-TKBS-0033 Revision No. 01 Effective Date: 10/16/2006 Page 18 of 32

exposure to the fire, or these last scenarios can be used to supplement intakes for workers whose 

bioassays are only associated with the fire. The dose distribution is assumed to be constant.

 Table 3-3. Inhalation intake summary for operational period. 

Scenariosa Start End

Intake 

type

Absorption 

type

Intake 

(pCi/day or pCi)

Fire plus workplace exposure 1/1/1948 4/30/1949 Chronic M 4.81E+02

6/10/1948 -

b Acute M 2.17E+03

Fire plus workplace exposure 1/1/1948 4/30/1949 Chronic S 1.43E+04

6/10/1948 -

b Acute S 7.39E+04

Workplace exposure 1/1/1948 4/30/1949 Chronic M 4.96E+02

Workplace exposure 1/1/1948 4/30/1949 Chronic S 1.49E+04

a. Only one of the four scenarios from the table is used to calculate an organ dose. The scenario choice depends on 

whether the worker could have been exposed to the fire between June 1 and June 11, 1948. The choice also depends 

on the organ of interest.

b. Not applicable.

4.0

Individual external dosimetry results for Chapman Valve are reported for the weeks beginning May 3

to November 1, 1948 (AEC 1948b). Exposures reported for the weeks beginning October 11, 18, and 

25 are questionable, because one set of data indicates that no results were reported and another set 

indicates that results were all less than 50 milliroentgen (mR). This apparent discrepancy may be due 

to the practice of expediting data entry by marking multiple results that were “less than” with an x. 

When a whole page of results were “less than”, the entire result section appears to have been marked 

with a large x. Lines marked with the word “None”, indicate a badge was not worn. Reporting of 

numerical results on an x-marked page indicates positive results and these should be included in the 

record.

ESTIMATION OF EXTERNAL EXPOSURE

Because film badge data are available for Chapman Valve workers, this document does not attempt 

to address worker external exposures based on workplace data. When film badge results are 

available for a worker, the individual’s dosimeter results can be used to estimate dose. This 

document also provides an upper estimate of external dose based on film badge dosimetry records.

The majority of photons from natural uranium metals are in the 30- to 250-keV energy range. Solid 

uranium objects provide considerable attenuation of the lower energy photons and harden the 

spectrum, causing the majority of photons emitted from a solid uranium object, such as a billet or a 

rod, to have energies greater than 250 keV. While it is recognized that solid uranium sources have a 

hardened photon spectrum, exposure to a thin layer of uranium on a surface results in a larger 

fraction of exposure to lower energy photons. This analysis assumed workers were exposed to 

photon energies in the 30- to 250-keV range, which is favorable to claimants. Nonpenetrating dose 

from natural uranium consists primarily of electrons with energies >15 keV. For consistent 

presentation, exposure or dose is reported as:

• Penetrating, assumed to be associated with photons of energies 30 keV or greater, and

• Nonpenetrating, assumed to be associated with photons of energies less than 30 keV or with 

electrons.
Document No. ORAUT-TKBS-0033 Revision No. 01 Effective Date: 10/16/2006 Page 19 of 32

4.1 SUMMARY OF AVAILABLE FILM BADGE DATA

Chapman Valve (Musgrave 1948) reported the following ancillary information regarding film badge 

availability and use at Chapman Valve: 

• February 2, 1948: received 50 (brass) badges from the University of Rochester. 

• February 9, 1948: received duplicate shipment (with identical numbers) noted to be the pintype. 

• April 28, 1948: requested 100 badges and subsequently received 100 stainless steel snap-on 

badges from University of Rochester.

• May 10, 1948: returned the 50 used brass pin-type dosimeters and 50 used stainless steel 

snap-on badges to University of Rochester.

• May 18, 1948: received 100 stainless steel snap-on badges and noted that Chapman Valve 

would like to continue with this badge type.

• May 19, 1948: sent 50 brass pin-type and 50 stainless steel snap-on badges to AEC in New 

York per AEC request.

Less than 50 film badge results are included in each reporting period, most likely indicating that only 

some of the workers entered the AEC work area in a given week (AEC 1948b). A note at the bottom 

of some of the May reports states “Eastman Film” but does not mention the type of film. Exposure for 

both beta and gamma was reported in “mr” in 1948. The minimum reporting limit was 50 mR (though 

at least one report mistakenly lists the reporting limit as “0.50 mr”) for both beta and gamma. Some 

reports include names and some do not. Some external dosimetry report copies contain illegible 

entries. As of this writing, it has not been determined if a complete data set for a worker can be 

determined from the multiple copies of dosimetry reports. An entry of “None” is included on some 

reports, and this seems to indicate that the badge was not worn during the monitoring period. The 

handwritten response to the question, “Where were badges kept overnight?” was “No,” which 

probably indicates that workers left their badges on site at the end of the workday. There is no 

indication of where the badges were stored or that a control badge was used.

Because it might not be possible to associate results with individuals because of the poor copies or 

the censored identifiers, the data have been reviewed to identify the maximum recorded beta and 

gamma doses for each week. The median beta result for May (115 mR) and the minimum reporting 

limit specified on the Chapman Valve reports for gamma (50 mR) were used to estimate the exposure 

for other weeks when film badges were not in use or for weeks when the data were suspect, e.g., the 

three noted weeks in 1948 beginning October 11, 18, and 25. Exposures during unmonitored weeks 

were likely to be lower, because according to records, the majority of processing occurred between 

May and November. Some of the results for the weeks of May 17 and 24, 1948 were combined in 

some of the original records. No attempt was made to unfold the combined results; instead, the 

maximum reported results for May 17 and 24, 1948, and the May 24, 1948 report were assumed to be 

the maximum results for the week of May 17 and the week of May 24, respectively. For the weeks 

May 10 to September 27, 1948, the gamma reporting limit (0.05 R) was used to estimate the 

maximum exposure when it was reported as <0.05 R. For the weeks May 31 to June 14, 1948, 

inclusive, the maximum result for the three weeks was assumed for each of these weeks, because of 

the uncertainty in determining which result applied to which week. When the period of the exposure is 

less than that included in Table 4-1, the exposure should not necessarily be reduced by the ratio of 

the actual to the total exposure period. Because recorded maximum external exposures were larger 

during the weeks beginning May 10 through October 4, 1948, this needs to be taken into account 

when estimating doses for shorter exposure periods.
Document No. ORAUT-TKBS-0033 Revision No. 01 Effective Date: 10/16/2006 Page 20 of 32

 Table 4-1. Maximum film badge results.a

Week

Beginning

Nonpenetrating

R

Penetrating

R

Week

Beginning

Nonpenetrating

R

Penetrating

R

1/5/1948 0.115 0.050 7/5/1948 0.225 0.050

1/12/1948 0.115 0.050 7/12/1948 0.240 0.050

1/19/1948 0.115 0.050 7/19/1948 0.260 0.050

1/26/1948 0.115 0.050 7/26/1948 0.320 0.050

2/2/1948 0.115 0.050 8/2/1948 0.240 0.085

2/9/1948 0.115 0.050 8/9/1948 0.360 0.110

2/16/1948 0.115 0.050 8/16/1948 0.260 0.085

2/23/1948 0.115 0.050 8/23/1948 0.260 0.050

3/1/1948 0.115 0.050 8/30/1948 0.260 0.050

3/8/1948 0.115 0.050 9/6/1948 0.440 0.070

3/15/1948 0.115 0.050 9/13/1948 0.500 0.075

3/22/1948 0.115 0.050 9/20/1948 0.650 0.050

3/29/1948 0.115 0.050 9/27/1948 0.320 0.050

4/5/1948 0.115 0.050 10/4/1948 0.160 0.055

4/12/1948 0.115 0.050 10/11/1948 0.115 0.050

4/19/1948 0.115 0.050 10/18/1948 0.115 0.050

4/26/1948 0.115 0.050 10/25/1948 0.115 0.050

5/3/1948 0.065 0.070 11/1/1948 0.085 0.070

5/10/1948 0.100 0.050 11/8/1948 0.115 0.050

5/17/1948 0.115 0.050 11/15/1948 0.115 0.050

5/24/1948 0.140 0.050 11/22/1948 0.115 0.050

5/31/1948 0.165 0.050 11/29/1948 0.115 0.050

6/7/1948 0.165 0.050 12/6/1948 0.115 0.050

6/14/1948 0.165 0.050 12/13/1948 0.115 0.050

6/21/1948 0.130 0.060 12/20/1948 0.115 0.050

6/28/1948 0.265 0.050 12/27/1948 0.115 0.050

Totals 9.110 2.830

a. Numbers in italics are assumed. The bases for the assumptions are presented in the text above.

An additional period of external exposure is assumed from January 3, 1949 through May 1, 1949. 

The median weekly beta result for May 1948 (0.115 R) and the gamma reporting limit of 0.050 mrem 

were used to estimate exposure during this period. The results are shown in Table 4-2.

4.2 OCCUPATIONALLY REQUIRED MEDICAL X-RAY

Information regarding whether or not occupationally required medical X-ray examinations were 

performed at Chapman Valve is unavailable. AEC usually, but not always, required preemployment 

and periodic medical examinations of workers involved in the larger uranium processing programs. 

The term preemployment as used here means prior to performing AEC-contracted radiological work. 

The typical AEC medical program included preliminary and annual chest X-ray examinations. This 

analysis assumed that workers received a preemployment X-ray examination of the chest in 1947 and 

a second X-ray examination a year later. The method of X-ray examination should be based on the 

current guidance for 1948 exposures. Organ doses can be obtained from the latest revision of 

ORAUT-OTIB-0006, Technical Information Bulletin: Dose Reconstruction from Occupationally Related 

Diagnostic X-Ray Procedures (ORAUT c).
Document No. ORAUT-TKBS-0033 Revision No. 01 Effective Date: 10/16/2006 Page 21 of 32

4.3 MISCELLANEOUS INFORMATION RELATED TO EXTERNAL DOSE

This section includes external dose information that might be of interest for specific dose 

reconstructions. This analysis did not consider such information generically because of its limited 

applicability or because of limited information.

A fire involving uranium occurred sometime between June 1 and 11, 1948. The fire and subsequent 

cleanup activities could have increased the likelihood of uranium skin contamination on some 

workers. Film badges were worn by Chapman Valve workers during this period, but it has not yet 

been verified that the workers who fought the fire or were involved in cleanup were monitored for 

external radiation exposure. 

4.4 OCCUPATIONAL EXTERNAL DOSE RECONSTRUCTION ASSUMPTIONS AND 

SUMMARY

Individual film badge results are available to determine doses. The reporting limit for beta and gamma 

was 50 mR. Table 4-2 provides overestimating assumptions that can be used to estimate doses for 

some Chapman Valve claims including Interactive RadioEpidemiological Program (IREP) distribution 

information. Prorating of the exposure to a shorter period should be based on the information in Table 

4-1. 

 Table 4-2. External exposure summary. 

Exposure 

category Exposure type Basis Year

Annual 

exposure IREP distribution

Overestimate 

of external 

dose

Penetrating Maximum film badge 

results

1948

1949

2.830 R

0.850 R

Constant

Nonpenetrating Maximum film badge 

results

1948

1949

9.110 R

1.955 R

Constant

Medical X-ray Current guidance 

for 1947 and 1948

Initial exam in 1947 

plus one exam in 1948

1947

1948

See ORAUT-OTIB-0006, (ORAUT

2005c)

5.0

The residual dose period is assumed to begin on May 1, 1949, the day after the month that Electro 

Metallurgical (1949) noted processing of Chapman Valve material and scraps, and is assumed to 

continue through December 31, 1993. The radiation exposures during cleanup operations in 1994 

and 1995 are not assessed in this document.

ESTIMATION OF RESIDUAL EXPOSURE

In 1987, Department 40 and its perimeter were surveyed with a Victoreen Model 492 ionization 

chamber, and no elevated radiation levels were detected (Sedelow 1987). The detection threshold of 

a Victoreen 492 is likely to be about 0.1 mR/hour. 

ORNL (1992) conducted a survey in August 1992 under the FUSRAP program. The reported 

radiation levels at various locations in the facility are summarized in Table 5-1. Elevated radiation and 

contamination levels were found in the western portion of Building 23. The survey of Chapman Valve 

included:

• a gamma and beta/gamma scan of the floor and walls and gamma scan of the ground surface 

in selected outdoor areas.

• Measurement of surface and 1-m exposure rates at the center of the north and south section 

of main-bay survey blocks. 
Document No. ORAUT-TKBS-0033 Revision No. 01 Effective Date: 10/16/2006 Page 22 of 32

• Measurement of alpha activity levels at selected locations. 

• Analysis of 30 dust and debris samples (26 from overhead beams). 

• Direct and removable alpha and beta/gamma measurements at 31 locations.

• Analysis of 2 soil samples.

 Table 5-1. Radiation/radioactivity levels in August 1991.a

Measurement type Range MDA

Gamma exposure (μR/hr) 5–32 --b

Direct beta/gamma (mrad/hr) 0.02–4 0.01

Direct alpha (dpm/100 cm2) <MDA –2,900 25

Removable beta/gamma (dpm/100 cm2) all <200 200

Removable alpha (dpm/100 cm2) 14–90 10

Soil concentrations (pCi/g)c 0.33–1.9 --

Dust samples (pCi/g) c 0.2–36,000 --

a. ORNL (1992). 

b. -- = not available. 

c. Includes Ra-226, Th-232, and U-238.

All results in the ORNL reports, except those from the smears, were reported as gross results; 

background values were not subtracted from the results. ORNL (1992) included brief information 

regarding their survey methods and instrumentation.

A sodium iodide scintillation probe and rate meter were used to detect gamma. Rates measured at 

2 inches from a surface were converted to μR/hr. The method to measure the exposures at 1 meter, 

which are included in the report, was not explicitly stated. Gamma levels ranged from 4 to 32 μR/hr. 

The highest average reading for a grid block was 13 μR/hr. It was assumed that these measurements 

were made at 1 meter from the floor. To estimate the penetrating dose it was assumed the largest 

average result, 13 μR/hr, was the median and the largest result, 32 μR/hr, was at the 95th percentile. 

An associated geometric standard deviation (GSD) of 1.72 was calculated. The penetrating dose 

estimates are listed in Table 5-2. 

A pancake Geiger-Mueller (GM) probe measured count rates, and these were converted to mrad/hr of 

beta/gamma. The minimum detectable activity (MDA) was reported as 0.01 mrad/hr. A pancake GM 

probe is an unusual choice for making dose measurements because of both its geometry and energy 

sensitivities. It is also sensitive to alpha radiation, which might have resulted in an over-response if a 

cover or some distance from a surface were not maintained to eliminate the alpha response. ORNL 

(1992, Figure 7) reported a range of dose rates for each grid block surveyed. The maximum dose 

rates in the range were used to determine a geometric mean, which was assumed to equal a median 

nonpenetrating dose rate of 0.12 mrad/hr. The GSD was calculated to be 2.9. The nonpenetrating 

dose is listed in Table 5-2. 

A zinc sulfide scintillation detector and rate meter were used to detect alpha count rates. The count 

rates were converted to dpm/100 cm2

. The MDA was reported as 25 dpm/100 cm2. The largest of 

30 direct contamination measurements was reported as 2,900 dpm/100 cm2

. To calculate internal 

exposure from residual activity, this analysis assumed that the median uranium exposure was 

associated with uniform contamination of the buildings to a level of 2,900 dpm/100 cm2

, the maximum 

directly measured alpha contamination. Six of 23 dust samples from the overhead beams and crane 

exceeded this value and ranged from 3,500 dpm to 12,000 dpm/100 cm2

. Using a resuspension 

factor of 1 × 10-6/m (NRC 2002) and an air intake rate of 2,400 m3 per work year, the calculated 

annual inhalation intake was 314 pCi, and this was assumed to all be uranium. Using the method 

described in NIOSH (2004), the calculated annual ingestion intake was 6.53 pCi. GSDs of 3 are 

assumed. Table 5-2 summarizes residual period intake rates.
Document No. ORAUT-TKBS-0033 Revision No. 01 Effective Date: 10/16/2006 Page 23 of 32

Samples of dust and debris, and soil were analyzed for Ra-226, Th-232 and U-238 by ORNL in the 

1990s (ORNL 1992, 1997). Chapman Valve received essentially pure uranium metal (no radium) for 

processing. This is confirmed by the ORNL (1992) survey, which reported Ra-226 and Th-232 

concentrations as consistent with background. The two soil samples were also consistent with 

background radioactivity concentrations. ORNL analyzed dust samples M10 and M31 for U-235. 

M31 was reported as 2.16% enriched. The Chapman Valve uranium processing was related to 

reactor needs, and the use of enriched uranium in reactors was rare until 1950. The only other 

mention of Chapman Valve and enriched uranium was in ORNL (1997), which reiterated ORNL’s 

1992 words. Unlike the ORNL 1992 report, ORNL 1997 included the actual U-235 and U-238 results 

for seven soil samples. The activity ratios seemed consistent with natural uranium. However,

because the concentrations were close to background, the reported uncertainties were relatively 

large. 

The estimated annual external exposures to residual radioactivity from AEC operations at the site, 

listed in Table 5-2, were calculated by assuming that workers were exposed for 2,000 hours per year. 

Assumptions regarding residual exposures should be consistent with assumptions from the 

operational period.

 Table 5-2. Annual internal and external exposure to residual radioactivity. 

Internal Start End Exposure 

Absorption 

type

Intake 

(pCi/day) IREP distribution

Uranium 5/1/1949 12/31/1993 Inhalation M, S 8.58E-01 Lognormal GSD 3

5/1/1949 12/31/1993 Ingestion (a) 1.79E-02 Lognormal GSD 3

External Start End Exposure Basis R/year IREP distribution

5/1/1949 12/31/1949 Penetrating Survey instrument 1.73E-2 Lognormal GSD 1.72

1/1/1950 12/31/1993 Penetrating Survey instrument 2.60E-2 Lognormal GSD 1.72

5/1/1949 12/31/1949 Nonpenetrating Survey instrument 1.57E-1 Lognormal GSD 2.9

1/1/1950 12/31/1993 Nonpenetrating Survey instrument 2.36E-1 Lognormal GSD 2.9

a. Choose same f1-value as used for inhalation per NIOSH (2004).
Document No. ORAUT-TKBS-0033 Revision No. 01 Effective Date: 10/16/2006 Page 24 of 32

REFERENCES

AEC (U.S. Atomic Energy Commission), 1948a, Monthly Status and Progress Report for October 

1948, Office of New York Directed Operations, November 14. [SRDB Ref ID: 6287; contains 

Export Controlled Information] 

AEC (U.S. Atomic Energy Commission), 1948b, NYDO Radiological Laboratory, U.S.A.E.C. Film 

Badge Report, Plant, Chapman Valve Mfg. Co., U.S. Atomic Energy Commission, New York, 

New York. [pp. 5-177 of Chapman Valve_1948.pdf]

AEC (U.S. Atomic Energy Commission), 1948c, AEC-NYDO Sample Record # _, Radiological 

Laboratory, New York, New York, May 6. [p. 1 of EML2Bx22ChapmanValveDust1948.pdf]

AEC (U.S. Atomic Energy Commission), 1948d, AEC-NYDO Sample Record # 4, Radiological 

Laboratory, New York, New York, June 1. [p 2 of EML2Bx22ChapmanValveDust1948.pdf]

AEC (U.S. Atomic Energy Commission), 1949, Health Hazards in NYOO Facilities Producing and 

Processing Uranium: A Status Report, April 1, 1949, New York Operations Office, Medical 

Division, New York, New York. [SRDB Ref ID: 17068]

Author unknown, 1948, Uranium Content of Urine Samples Collected at Chapman Valve Mfg. 

Company. [pp. 4-6 of 002063.pdf]

Author unknown, ca. 1996, “Figure 2, Chapman Valve, Building 23 – Plan View.” [SRDB Ref ID: 

25012]

BNL (Brookhaven National Laboratory), 2006, “Brookhaven Graphite Research Reactor,”

http://www.bnl.gov/bnlweb/history/BGRR.asp, accessed September 14. 

Eidson, A. F., 1994, “The Effect of Solubility on Inhaled Uranium Compound Clearance: A Review,” 

Health Physics, volume 67, number 1, pp. 1-14. 

Eisenbud, M.,1948, “Dust and Water Samples,” letter to T. P. Musgrove (Chapman Valve Company), 

U. S. Atomic Energy Commission, New York Operations Office, May 11. [p. 3 of 

EML2Bx22ChapmanValveDust1948.pdf]

Electro Metallurgical, 1949, Chapman Valve Account, Electro Metallurgical Division U. C. & C. Corp. – 

Area Plant, Weekly Production Report, April 1 to April 30. [pp. 8 and 24 of Electro 

Metallurgical.pdf]

Fiore, J. J., 1987, untitled letter to P. Hundt (The Crane Company), U.S. Department of Energy, 

September 29. [pp. 13–16 of Chapman Valve.pdf]

Fox, G. E., 1949, untitled letter to B. S. Wolf, (U.S. Atomic Energy Commission), Chapman Valve 

Manufacturing Company, Indian Orchard, Massachusetts, January 19. [030000686.pdf]

Harris, W.B., and I. Kingsley, 1959, The Industrial Hygiene of Uranium Fabrication, Table 5, Archives 

of Industrial Health, volume 19, reprint, American Medical Association, New York, New York, 

May. [SRDB Ref ID: 15779]
Document No. ORAUT-TKBS-0033 Revision No. 01 Effective Date: 10/16/2006 Page 25 of 32

Heffernan, T. E., J. C. Lodwick, H. Spitz, J. Neton, and M. Soldano, 2001, “Solubility of Airborne 

Uranium Compounds at the Fernald Environmental Management Project,” Health Physics, 

volume 80, number 3, pp. 255-262.

ICRP (International Commission on Radiological Protection), 1975, Report of the Task Group on 

Reference Man, Publication 23, Pergamon Press, Oxford, England.

ICRP (International Commission on Radiological Protection), 1994, Human Respiratory Tract Model 

for Radiological Protection, Publication 66, Pergamon Press, Oxford, England.

ICRP (International Commission on Radiological Protection), 1995, Age-dependent Doses to 

Members of the Public from Intake of Radionuclides: Part 4 Inhalation Dose Coefficients, 

Publication 71, Pergamon Press, Oxford, England.

LeVine, H. D., 1948, “Dust Sample Results,” letter to V. T. Malcolm, (Chapman Valve Manufacturing 

Company), Radiological Laboratory, March 19. [p. 5 of 

EML2Bx22ChapmanValveDust1948.pdf]

Morgan, J. P., 1948, “Accounting for Source and Fissionable Material, ” letter to G. E. Fox (Chapman 

Valve Manufacturing Company), U. S. Atomic Energy Commission, New York Operations

Office, August 24. [Chapman Valve Material.pdf]

Morgan, J. P., 1949, “SF Material Reports,” letter to G. E. Fox (Chapman Valve Manufacturing 

Company), U. S. Atomic Energy Commission, New York Operations Office, January 25. [pp. 

10-11 of Chapman Valve.pdf]

Musgrave, T. P., 1948, “Film Exposure Badges, Chapman Valve Mfg. Co. Indian Orchard, Mass.,” 

letter to H. D. LeVine (U.S. Atomic Energy Commission), H.K. Ferguson Company at the 

Chapman Valve Mfg. Co., May 24. [p. 4 of Chapman Valve_1948.pdf]

NIOSH (National Institute for Occupational Safety and Health), 2004, Estimation of Ingestion Intakes, 

Technical Information Bulletin, OCAS-TIB-009, Rev. 0, Office of Compensation Analysis and 

Support, Cincinnati, Ohio, April 13. 

NRC (U.S. Nuclear Regulatory Commission), 2002, Re-evaluation of the Indoor Resuspension Factor 

for the Screening Analysis of the Building Occupancy Scenario for NRC’s License Termination 

Rule, NUREG-1720; Draft Report for Comment, Washington, D.C. [SRDB Ref ID: 22400]

ORAUT (Oak Ridge Associated Universities Team), 2005a, February 14, 2005 Meeting with:

Chapman Valve Manufacturing, held at Western Massachusetts COSH office, Springfield,

Massachusetts, Final Minutes, Oak Ridge, Tennessee, September 2. 

ORAUT (Oak Ridge Associated Universities Team), 2005b, Internal Dosimetry Coworker Data for 

Y-12, ORAUT-OTIB-0029, Rev. 00, Oak Ridge, Tennessee, April 5.

ORAUT (Oak Ridge Associated Universities Team), 2005c, Dose Reconstruction from Occupationally 

Related Diagnostic X-Ray Procedures, ORAUT-OTIB-0006, Rev. 03 PC-1, Oak Ridge, 

Tennessee, December 21. 
Document No. ORAUT-TKBS-0033 Revision No. 01 Effective Date: 10/16/2006 Page 26 of 32

ORAUT (Oak Ridge Associated Universities Team), 2006, SEC Petition Evaluation Report, Petition 

SEC-00043, Rev # 0, Oak Ridge, Tennessee, August 8. 

[http://www.cdc.gov/niosh/ocas/pdfs/sec/chapman/chapmevalr.pdf] 

ORNL (Oak Ridge National Laboratory), 1992, Results of the Radiological Survey at the Former 

Chapman Valve Manufacturing Company, Indian Orchard, Massachusetts, ORNL/RASA-92/1, 

Environmental Restoration and Waste Management Non-Defense Programs, Oak Ridge, 

Tennessee. [Chapman Valve - ORNL Rad. Survey.pdf]

ORNL (Oak Ridge National Laboratory), 1997, Results of the Independent Radiological Verification 

Survey at the Former Chapman Valve Manufacturing Company, Indian Orchard, 

Massachusetts (CIO001V), ORNL/RASA-95/17, Environmental Restoration and Waste 

Management Non-Defense Programs, Oak Ridge, Tennessee, May. [Chapman Valve Rad. 

Verification Survey.pdf]

ORNL (Oak Ridge National Laboratory), 2003, Rad Toolbox v. 1.0.0, software, December 31, 

http://ordose.ornl.gov/downloads.html. 

Redacted, 1987, “Crane – Indian Orchard,” letter to J. J. Fiore (U.S. Department of Energy), Crane 

Company, New York, New York, December 14. [pp. 1-9 of Chapman Valve U Machining.pdf]

Robbins, R., 1996, “Chapman Valve Final Remedial Action Report,” memorandum to M. Redmon, 

Bechtel FUSRAP Project Job 14501, May 13. [SRDB Ref ID: 13374]

Schnell, E., and R. Caldwell, 1967, “Incinerator,” memorandum to T. Morton and D. Gabrielli, NUMEC 

[handwritten on document], August 18. [SRDB Ref ID: 23778]

Sedelow, C. A., 1987, untitled letter to Crane Company, Crane Valve Services, Indian Orchard, 

Massachusetts, November 19. [p. 5 of 030000808.pdf]

Springfield Union News, ca. 1942, “Chapman Valve Resumes Work This Morning,” Springfield, 

Massachusetts, pp. 1 and 9. [SRDB Ref ID: 15563]

Springfield Union News, ca. 1952, “Chapman Valve Atomic Role Revealed,” Springfield, 

Massachusetts. [SRDB Ref ID: 15563]

Williams, W. J., date unknown, memorandum (excerpt) to W. E. Kelley (Manager, New York), “SF 

Materials Lost or Unaccounted for,” Director of Production, Washington. [pdf pp. 39-40 of 

http://www.cdc.gov/niosh/ocas/pdfs/sec/chapman/chapmpet.pdf] 

Wolf, B. S., 1947, “Monthly Report,” memorandum to H. B. Fry, AECM-11 NY, U.S. Atomic Energy 

Commission. [p. 1 of Chapman Valve Monthly Report Survey.pdf]

Wolf, B. S., 1948a, memorandum to C. Slesser (Publications Branch), “Weekly Activity Report – 

Period Jan. 15 to Jan. 21,” memorandum to C. Slesser, AECM-11 NY, U.S. Atomic Energy 

Commission, January 21.. [p. 2 of Chapman Valve Weekly Activity Report.pdf] 

Wolf, B. S., 1948b, “Weekly Activity Report – April 28 to May 5,” memorandum to C. Slesser, AECM11 NY, U.S. Atomic Energy Commission, May 6. [p. 5 of Chapman Valve Weekly Activity 

Report.pdf]
Document No. ORAUT-TKBS-0033 Revision No. 01 Effective Date: 10/16/2006 Page 27 of 32

Wolf, B. S., 1949, untitled letter to G. E. Fox (Chapman Valve Manufacturing Company), U.S. Atomic 

Energy Commission, January 27. [p. 5 of 030000686.pdf]

Young, C. D., 1987, “Request for Information – Crane Company Sites,” letter to A. Wallo, III (U.S. 

Department of Energy), The Aerospace Corporation, Washington, D.C., August 31. [p. 2-8 of 

Chapman Valve.pdf]
Document No. ORAUT-TKBS-0033 Revision No. 01 Effective Date: 10/16/2006 Page 28 of 32

ATTACHMENT A 

CHAPMAN VALVE BUILDING LAYOUT AND SOME SAMPLING LOCATIONS

Page 1 of 3

Figure A-1. Chapman Valve floor plan provided at 2005 Worker Outreach Meeting (Author unknown, ca 1996). 
Document No. ORAUT-TKBS-0033 Revision No. 01 Effective Date: 10/16/2006 Page 29 of 32

ATTACHMENT A 

CHAPMAN VALVE BUILDING LAYOUT AND SOME SAMPLING LOCATIONS

Page 2 of 3

Figure A-2. Locations of dust and debris samples collected in 1992 in or near Building 23 (ORNL 1992).
Document No. ORAUT-TKBS-0033 Revision No. 01 Effective Date: 10/16/2006 Page 30 of 32

ATTACHMENT A 

CHAPMAN VALVE BUILDING LAYOUT AND SOME SAMPLING LOCATIONS

Page 3 of 3

Figure A-3. Locations of verification samples collected in 1996 in Building 23 (ORNL 1997). 
Document No. ORAUT-TKBS-0033 Revision No. 01 Effective Date: 10/16/2006 Page 31 of 32

ATTACHMENT B 

COMPARISON OF AIR CONCENTRATIONS CALCULATED FOR CHAPMAN VALVE AND AEC 

MACHINING AND GRINDING OPERATIONS

Page 1 of 2

Harris and Kingsley (1959) summarized uranium operational data in an article submitted for 

publication to the American Medical Association Archives of Industrial Health for June 6, 1958, (which 

was previously issued as AEC Health and Safety Laboratory (HASL) report HASL-39). The data 

appeared to be collected from several uranium metal handling facilities, which probably included 

Fernald, Mallinckrodt Chemical Works, Simonds Saw and Steel, Bridgeport Brass Company, and 

Sylvania-Corning Nuclear (Harris and Kingsley 1959, p. 113). Based on the references and the 

involved companies, it is likely that the data were collected before 1957 and perhaps as early as the 

late 1940s. Harris and Kingsley noted that the information was collected from both full-time 

operations at production plants, as well as from experimental runs on “production-sized” units, and 

included collection of data for both controlled ventilation and no mechanical ventilation conditions. 

The air concentrations reported as daily averages were based on time-weighting the measured air 

concentration by each incremental exposure period incurred by the operator in a workday. These 

HASL-collected numbers are compared with the air concentration rates that would result in the intake 

rates derived from Chapman Valve bioassay. For the non-fire scenarios, these air concentrations 

were calculated by multiplying the intake rate in pCi per day by 2.22 dpm/pCi, multiplying that result 

by 365 days per year and dividing it by 2,000 work-hours/year and the light-work breathing rate of 1.2 

m3

/hr (ICRP 1995) to get the air concentration in dpm/m3. To include the fire scenario, total intakes 

for the operational period were calculated and then averaged to a daily intake rate in pCi per day 

before calculating average air concentrations as stated above.

Table B-1 lists the calculated Chapman Valve air concentrations and the reported air concentrations 

from Harris and Kingsley (1959) and shows that the air concentrations derived for the Chapman Valve

machining operation tend to be on the high side of the machining operations presented by HASL. The 

operation at Chapman Valve appeared to be primarily a machining operation and not so much a 

cutting and grinding operation, but a comparison with the abrasive operations can also be made, and 

shows that the Chapman Valve calculated air concentrations would be in similar range. It is likely that 

machining and abrasive operations at Chapman Valve were ongoing for less than 8 hours a day and 

250 days a year during the period 1948 to April 30, 1949, and this gives assurance that the calculated 

air concentrations for Chapman Valve are bounding for this type of work.
Document No. ORAUT-TKBS-0033 Revision No. 01 Effective Date: 10/16/2006 Page 32 of 32

ATTACHMENT B 

COMPARISON OF AIR CONCENTRATIONS CALCULATED FOR CHAPMAN VALVE AND AEC 

MACHINING AND GRINDING OPERATIONS

Page 2 of 2

Table B-1. Comparison of air concentrations calculated for Chapman Valve and AEC machining and 

grinding operations (dpm/m3

). 

Chapman Valve

Machining operations 160–5,000

AEC facilitiesa

Daily averagesb Breathing zone General area

Ventilation

No 

ventilation Ventilation

No 

ventilation Ventilation

No 

ventilation

Machiningb

Lathing (optimal speed) --c -- <1 3–90 -- --

Lathing (cutting speed 

200–300 surface ft/min, 

coolant rates up to 115 

gallons/min)

-- -- -- -- 12–48 120–240

Automatic lathe 30–70 200–300 -- -- -- --

Turret lathed 40–50 150 -- 1,750 -- --

Facingd -- ≈100 -- -- -- --

Cutoffd 20–30 ≈100 -- -- -- --

Millingd 20–30 ≈100 -- -- -- --

Slottingd 20–30 ≈100 -- -- -- --

Drill 10 20 -- -- -- --

Radius cuttingd 30 100–300 -- -- -- --

Milling -- 40 -- -- -- --

Shaping -- <10 -- -- -- --

Planing -- <10 -- -- -- --

Abrasive operationse

Cut-off <1 --f -- -- -- --

Surface grinder 50–200 2,000–5,000 -- -- -- --

Portable grinder 50–200 400 -- -- -- --

Belt sander <10 3,000 -- -- -- --

Centerless grinder 50–300 5,000–6,000 30–270 13,500 -- --

a. Harris and Kingsley (1959). 

b. Daily averages are from Harris and Kingsley (1959, Table 5). 

c. Not provided. 

d. Values are for normal operations; with speeds of less than 100 surface ft/min and adequate coolant, all results are less 

than 10 dpm/m3

. 

e. Harris and Kingsley (1959, Table 6). 

f. “Never sampled, but very high” (Harris and Kingsley 1959).